The paper presents results of certification of the ATHLET/ BIPR-VVER software system (version 1.0) in the Russian Federation. The following stages of the software system certification are described: preparation of a validation report; work with appointed experts; performance of additional studies to increase the level of the software system validation; and preparation of a software system certificate. Examples of validation computations simulating experiments in operating power units and thermal-hydraulic experimental facilities are given. The ATHLET/BIPR-VVER software system (version 1.0) was certified by the Expert Council on software certification under Rostechnadzor on 24.10.2018 (Certificate No. 455).

The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre “Kurchatov Institute” (with the code ATHLET/BIPR-VVER), ÚJV Řež (with the code RELAP5-3D©) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

A brief description of the development history of engineering computer codes for VVER reactor calculations, which have been developed in the VVER physics department of NRC Kurchatov institute, since the end of 1950 s till now, are given in the report. The modern status of the basic codes of the Kurchatov institute for VVER fuel loadings design and safety analysis (TVS-M, KASKAD-BIPR-PERMAK, BIPR8, PERMAK-3D, SC1, TOPRA, ATHLET/BIPR-VVER, TIGR-1, etc.) is characterized from various points of view. Some specific problems which are shown during designing of fuel cycles, optimisation of fuel and technologies will be commented. They demand decisions within the complex of programs and, accordingly, demand the further development of a complex or working out and inclusion in a complex of new programs.

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The modern coupled thermal-hydraulics neutron–kinetics code systems require validation by comparison with measured local core data. Measured assembly coolant temperatures at core outlet are applied for this purpose in this study. Experimental data have been collected during the commissioning phase of the NPP Kalinin-3. One of the experimental data sets is selected to be used for international benchmark purposes (1). The new modelling capabilities of the ATHLET/BIPR-VVER code as well as the validation on plant measurements are discussed in the paper. The detailed modelling features of the reactor core allow to predict the coolant temperature at the assembly outlet within a rather high accuracy even though ATHLET code is based on one dimensional thermal-hydraulic pipe models.